This is probably old news for people, but I caught this link about Thor Energy starting 5 year trials (in April 2013?) of thorium-MOX fuel rods in a demo reactor for steam process heat applications at a Norwegian paper mill. The thorium-MOX fuelrod seems to be intended to be a thorium and plutonium fuel rod for LWR's, burning plutonium waste in likely ABWR variants.

The Halden research reactor is far from a demo reactor. It is a material test reactor that is quite known for determining solid oxide fuel irradiation limits. It is far from a demo reactor, you can find more information here: http://www.emtr.eu/hr.html

Nonetheless, this is an important step for any thorium fuel, because irradiation data is necessary before it may be loaded in any commercially operating plant.

_________________Liking All Nuclear Systems, But Looking At Them Through Dark And Critical Glasses.

http://dae.nic.in/writereaddata/.pdf_38Th-Pu MOX would, in any type of rector,1. Have a higher burn up and produce more power than U-Pu MOX.2. Have U-233 in the SNF.3. Dispose off plutonium.It is good that someone is trying it out in a reactor.

It won't necessarily produce higher burn-ups, you will get significant Protactinium losses due to the far higher equilibrium concentration in the fuel than the equivalent 239Np.

I have wondered about this quite a bit since I thought about fueling a CANDU with a 20% LEU 'driver' dispersed homogenously in thorium as a CANDU-SEU fuel.6% 20% enriched uranium and 94% thorium.

Good point, but one question is what is the current limit on burnup. It seems to be mostly cladding integrity from fission gas release and fission product attack. ThO2 matrix fuel solves this problem by having a lower fission gas release (higher melt point plus slightly higher thermal conductivity improves fuel performance).

Pu is attractive because of its inherently high "enrichment" in fissile. Much better than 20% LEU. And Pu destruction is very good with ThO2 compared to Pu-UO2 mox

http://dae.nic.in/writereaddata/.pdf_38Th-Pu MOX would, in any type of rector,1. Have a higher burn up and produce more power than U-Pu MOX.2. Have U-233 in the SNF.3. Dispose off plutonium.It is good that someone is trying it out in a reactor.

The question was if SNF from Th-Pu MOX would make the best LFTR startup fuel (except for pure U-233, since highly enriched U-235 seems to be a proliferation no-no). The mix was 90% Th-232, 10% Pu (don't know the Pu isotopic concentrations).

I think the answer is yes, please correct the logic below.

So essentially, unless the reactor burning Th-Pu MOX stalls (for lack of neutrons), it's continually fissioning pu-239, consuming more pu-240 than it makes, breeding u-233 (some gets fissioned), u-234 and u-235 (some gets fissioned), producing some Am and Cu, so in the end, we have mostly Th-232, Pa-233, U-233, U-234, U-235, a little unburned Pu, a little Am/Cu (considering it starts with just 10% Pu), and fission products.U-233 and U-235 are perfect for LFTR startup.If the fuel gets cooled for 10 months 99,9% of Pa-233 decays to more U-233.U-234 is better than Th-232 and much better than U-238 (just one extra neutron and it's U-235, while Th-232 has the Pa-233 decay delay and U-238 becomes Pu-239 upon neutron capture which has the 1/3 chance of becoming Pu240 instead of fissioning).Using LFTR planned capabilities, Th-Pu SNF would be fluorinated, undergo reprocessing for fission product removal, inserted into the volatility processing facility, tetra fluorides would go into the blanket, hexa fluorides bubble get reduced and go into the core, so I'm assuming Th and Pa goes into the blanket and U/Pu/Am/Cu going into the core.The main question mark is how much U-234 vs fissile material would be in the SNF. It's mainly a question of total fissile inventory in the LWR reactor per GWt (the more fissile inventory per GWt, the lower the chances of making U-234). Given LWR / HWR requires far more fuel per GWt due to the requirement for Xe-135 tolerance, U-234 production doesn't seem that bad.

The French group has been pushing this sort of a plan. Starting up on SNF/Pu and then transitioning to Th/233U. For their reactor you can start with SNF/Pu (ignoring solubility limits) and then feed the reactor only thorium and eventually burn off virtually all of the plutonium. Theirs is a fast reactor though. It requires more fissile and you bump up against plutonium solubility limits before you can start with just thorium + SNF/Pu. The solubility goes up considerably with higher temperature so they are planning to go with a higher temperature and take on the materials issues that come with that.

In a thermal reactor, the plutonium isn't quite as good a fuel so it gets tougher to start a breeder with plutonium. One suggestion is to start with plutonium and then pull out the uranium and leave the unburnt plutonium for another day and start with a fresh batch of salt plus the uranium. This would work.

The third approach is to do a converter. In this approach, you start each batch with plutonium but then after decades you have to replace the salt.

The French group has been pushing this sort of a plan. Starting up on SNF/Pu and then transitioning to Th/233U. For their reactor you can start with SNF/Pu (ignoring solubility limits) and then feed the reactor only thorium and eventually burn off virtually all of the plutonium. Theirs is a fast reactor though. It requires more fissile and you bump up against plutonium solubility limits before you can start with just thorium + SNF/Pu. The solubility goes up considerably with higher temperature so they are planning to go with a higher temperature and take on the materials issues that come with that.

In a thermal reactor, the plutonium isn't quite as good a fuel so it gets tougher to start a breeder with plutonium. One suggestion is to start with plutonium and then pull out the uranium and leave the unburnt plutonium for another day and start with a fresh batch of salt plus the uranium. This would work.

The third approach is to do a converter. In this approach, you start each batch with plutonium but then after decades you have to replace the salt.

Lots and lots of choices.

Yet lots and lots of choices, some more expensive than others Lars.The only drawback from the Thor energy halden test, is we're still not running an MSR, but given the choice of migrating a significant share of LWR / HWR to 90% Th, 10% Pu, or staying on LEU MOX, the Thor plan seems like the best transition to eventually having boatloads of LFTR reactors.

The more the world produces 233U (without 238U in the fuel), the better for future Thorium molten salt reactors.

Imagine if all of the worlds LWR / HWR on countries with Pu stockpiles (or could be supplied the prepared fuel), with just 10% reactor grade Pu, and almost free Thorium, fuel procurement should be much cheaper, and gone are the uranium stockpile concerns. Better burnup, cheaper, more available fuel. Load all water cooled reactors of NATO partners, Brazil, Japan, Australia, South Korea with that mix, and in 3 years once the fuel gets removed, we have a lot of LFTR startup fuel.

If we could have no fast reactors, I'd prefer that. Leave UO2 based SNF for later mixing on LFTR reactors such that less than 3% of the core runs of fissile SNF reprocessed with LFTR capabilities (no removal of 238U from SNF).

Fast reactors require 10x more fissile inventory. Humm why do I really prefer not having that ?

Sorry for being and ass, I used to love the S-PRISM idea. Now I think it will be a hog of all the SNF that we could use to start LFTRs in the future. And if GE truly believed in the S-PRISM concept, they would fund the first one somewhere. They certainly have the money in the bank, but of course, they are instead fishing for the first few ones be funded as SNF burners with the electricity being a useful side effect.

http://www.dae.nic.in/writereaddata/.pdf_31http://www.dae.nic.in/writereaddata/.pdf_37The Advanced Heavy Water reactor has been designed to burn mainly thorium with the three fissile materials. The use of thorium in various existing types of reactors has been studied in http://dae.nic.in/writereaddata/.pdf_38The practical trials in progress are to be welcomed.AHWR with U-233 and Pu fuel can produce enough U-233 for reuse but it is not a breeder. Pu will have to be supplemented. Initial U-233 will be produced in the blanket of the fast reactors. This could be used in a LFTR too if decided upon.

Jagdish, do you know more about the status of the AHWR in India ? The PFBR, the Indian sodium-cooled fast reactor will come on-line this year. The fast reactors are stage 2 in the Indian nuclear program. The AHWRs are part of stage 3. Are there already plans and concrete steps to construct these ? Areva, Rosatom and Westinghouse have been busy with trying to sell PWRs to India lately, if I recall correctly.

Thanks, Jagdish, the article was very informative. The construction of the AHWR is slated to start in 2016 and the reactor is likely to be operational by 2025. I believe this reactor can also be run on PuO2/ThO2 fuel, which could make this very interesting for the UK, which wants to get rid of its 100 t Pu stockpile. The old colonial masters of India could supply India with the excess Pu that India needs to start its AHWRs. However, I don't know whether non-proliferation treaties would get in the way of such a plan.

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